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Fuel Element Experience at the Halden Boiling Water Reactor

Conference:

Abstract

The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was  More>>
Authors:
Aas, S.; [1]  Videm, K.; Hanevik, A. [2] 
  1. OECD Halden Reactor Project, Halden (Norway)
  2. Institutt for Atomenergi, Kjeller (Norway)
Publication Date:
Apr 15, 1968
Product Type:
Conference
Report Number:
IAEA-SM-99/6
Resource Relation:
Conference: Symposium on Heavy-Water Power Reactors, Vienna (Austria), 11-15 Sep 1967; Other Information: 9 refs., 9 figs., 1 tab.; Related Information: In: Heavy-Water Power Reactors. Proceedings of the Symposium on Heavy-Water Power Reactors| 1001 p.
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; ALUMINIUM; ALUMINIUM ALLOYS; BURNUP; CLADDING; ENRICHED URANIUM; FRACTURES; FUEL ELEMENTS; HBWR REACTOR; HEAT TRANSFER; HEAVY WATER; NATURAL URANIUM; NEUTRON FLUX; STAINLESS STEELS; STRESS CORROSION; URANIUM DIOXIDE; ZIRCALOY 2
OSTI ID:
22117099
Research Organizations:
International Atomic Energy Agency, Vienna (Austria)
Country of Origin:
IAEA
Language:
English
Other Identifying Numbers:
Other: ISSN 0074-1884; TRN: XA13M1051073720
Submitting Site:
INIS
Size:
page(s) 791-805
Announcement Date:
Aug 01, 2013

Conference:

Citation Formats

Aas, S., Videm, K., and Hanevik, A. Fuel Element Experience at the Halden Boiling Water Reactor. IAEA: N. p., 1968. Web.
Aas, S., Videm, K., & Hanevik, A. Fuel Element Experience at the Halden Boiling Water Reactor. IAEA.
Aas, S., Videm, K., and Hanevik, A. 1968. "Fuel Element Experience at the Halden Boiling Water Reactor." IAEA.
@misc{etde_22117099,
title = {Fuel Element Experience at the Halden Boiling Water Reactor}
author = {Aas, S., Videm, K., and Hanevik, A.}
abstractNote = {The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was installed in order to permit an increase in temperature to 230 Degree-Sign C and in power from 5 to 20 MW(th). The maximum burnup obtained is 11000 MWd/t and the maximum heat rating 375 W/cm with no fracture failure and practically no change in appearance according to the post-irradiation examination. One element was deliberately taken to burn-out conditions by throttling the water flow. After a series of burn-outs, the element finally failed because of over-temperature. The successful use of aluminium cladding at 150 Degree-Sign C mitiated an effort for making aluminium alloys suitable for normal power reactor operation. Promising properties were found for an alloy (designated IFA 3 aluminium) with A1 10% Si, 1% Ni, 1% Mg, 0.3% Fe + Ti. Despite increase in corrosion rate under heat transfer conditions, aluminium-clad elements have reached a bumup of 9000 MWd/t (av.) at a heat flux as high as 150 W/cm{sup 2}. (author)}
place = {IAEA}
year = {1968}
month = {Apr}
}